1. Field of the Invention
The present invention relates to a method for manufacturing zirconium-based alloys containing niobium for use in nuclear fuel rod cladding, wherein the manufactured alloy exhibits improved corrosion resistance.
2. Discussion of the Related Art
In the past, zirconium alloys have been widely applied to nuclear reactors, such as light water reactors and heavy water reactors. Such applications include nuclear fuel rod cladding, space grids, and reactor core components. Among zirconium alloys thus-far developed, Zircaloy-2 and Zircaloy-4 have been widely utilized. Here, Zircaloy-2 is comprised of 1.20-1.70 wt % of tin (Sn), 0.07-0.20 wt % of iron (Fe), 0.05-1.15 wt % of chromium (Cr), 0.03-0.08 wt % of nickel (Ni), 900-1500 ppm of oxygen (O), and the balance of zirconium (Zr); and zircaloy-4 is comprised of 1.20-1.70 wt % of tin, 0.18-0.24 wt % of iron, 0.07-1.13 wt % of chromium, less than 0.07 wt % of nickel, 900-1500 ppm of oxygen, and the balance of zirconium.
As the operating condition of nuclear power plants tends to be at high burnup, increased operating temperature, and high pH, Zircaloy-2 and Zircaloy-4 could not be utilized as nuclear fuel rod cladding. Recently, an extensive and successful research and development have been focused on increasing the corrosion resistance of zirconium-based alloys. A notable feature of the zirconium-based alloy developed in this manner is that the nuclear fuel rod cladding contains niobium to improve corrosion resistance.
The corrosion resistance of the zirconium-based alloy containing niobium depends on the alloying element, the size of precipitate in microstructure and the annealing condition. Particularly, the corrosion resistance of the zirconium-based alloy containing more than 1.0 wt % of niobium changes sensitively in the variation of niobium contents and annealing temperature. Therefore, to manufacture zirconium-based alloy containing niobium with superior corrosion resistance for use in nuclear fuel rod cladding, above all things, it is very important to establish an optimal manufacturing method.
The prior art related to the method for manufacturing Nb-containing zirconium alloys for use in nuclear fuel cladding tubes, space grids, and reactor core components is as following.
U.S. Pat. No. 5,838,753, EP Patent No. 895,247, 910,098, and 1,111,623, and JP Patent No. 11,109,072 disclose a method for manufacturing zirconium alloy comprising Nb (0.5-3.25 wt %) and Sn (0.3-1.8 wt %) for use in cladding tubes of high burn-up nuclear fuels. The method comprises heating the zirconium alloy billet at β range temperature above 950° C. and rapidly quenching the heated billet below a transformation temperature from (α+β) to α to form a martensitic structure; extruding the quenched billet below 600° C. to form a hollow billet; annealing the extruded billet below 590° C.; cold-working the annealed billet; and intermediate annealing to form nuclear fuel cladding tubes. As such, the nuclear fuel rod cladding tube has a microstructure in which second phase precipitates of β-Nb are distributed uniformly, intragranularly and intergranularly in the alloy matrix, thereby having a microstructure with excellent stability when irradiated by neutron.
WO Patent No. 2001-061062 discloses a method for manufacturing a nuclear fuel cladding tube comprising a low content of Sn and 0.60-2 wt % of Nb. Sn/Fe ratio is 0.25/0.5, 0.4/(0.35-0.5) or 0.5/(0.25-0.5). More than 0.75 wt % of Fe+Sn is added to the nuclear fuel cladding tube. The method is composed of vacuum melting, forging, and hot- and cold-rolling, followed by annealing. The thus-obtained alloy has β-Nb of small size and Zr—Nb—Fe precipitates uniformly distributed in the zirconium matrix.
JP Patent No. 2001-208879 discloses a nuclear fuel assembly composition comprising a welding part, wherein a zirconium alloy or a zircaloy alloy comprising 0.2-1.5 wt % of Nb is treated at a temperature of 400-620° C. to increase the corrosion resistance of the welding part.
WO Patent Nos. 2001-024193 and 2001-024194 disclose a zirconium alloy for reactor core components. The zirconium alloy comprises 0.02-1 wt % of Fe, 0.8-2.3 wt % of Nb, 2000 ppm or less of Sn, 2000 ppm or less of O, 5-35 ppm of S, and 0.25 wt % or less of Cr+V.
JP Patent No. 01-1158591 discloses a method for manufacturing zirconium alloy for a structure of reactor core or nuclear fuel cladding tube. The method consists of β-quenching, hot-working, cold-working, intermediate heat-treatment, final cold-working, and final annealing. At least one heat-treatment should comprise heating the zirconium alloy to above 750° C., and cooling it to 500° C. at a rate of about 40° C./s. Final annealing is then performed at 450-500° C.
JP Patent No. 06-049608 discloses a method for manufacturing zirconium alloy plate comprising steps of performing a solution heat-treatment, hot-working, heat-treatment, cold-working, and final annealing. The intermediate heat-treatment between the repeated cold-workings is performed once or several times so that the accumulated annealing parameter is limited to a range of 3×10−18 to 2×10−16. The temperature for hot-working is 700-800° C. and the annealing temperature is 400-650° C.
JP Patent No. 04-329855 discloses a method for manufacturing zirconium alloy comprising steps of melting Zr-2.5 wt % Nb alloy component, performing a solution heat-treatment at 870° C. for thirty minutes, water cooling, cold-working at a working percentage of 3.9%, and annealing at 500° C. for 24 hours. When the (α+β) type zirconium alloy is manufactured by a method comprising steps of performing a solution heat-treatment, cold-working at a working percentage of 1-5%, and annealing, the zirconium alloy exhibits high corrosion resistance and creep resistance.
JP Patent No. 63-050453 discloses a method for manufacturing zirconium alloy comprising steps of solution heat-treating a zirconium alloy comprising Nb, Sn, and Mo, cooling from (α+β) phase range or β phase range, cold-working, heating to a temperature higher than eutectoid temperature, cooling again, and annealing at α phase range lower than eutectoid temperature. The thus-obtained zirconium alloy exhibits properties of high strength and high corrosion resistance.
JP Patent No. 62-182258 discloses a method for manufacturing zirconium alloy comprising steps of solution heat-treating the zirconium alloy comprising 0.2-5 wt % of Nb, 0.5-3 wt % of Sn, 0.1-2 wt % of Fe, and 500-2000 ppm of O, at (α+β) phase range or β phase range; rapid cooling at a high rate; cold-working at an extrusion percentage above 15%; and annealing at a temperature higher than recrystallization temperature at the α phase range. The thus-obtained zirconium alloy exhibits excellent ductility and corrosion resistance. Also, the zirconium alloy comprising 2.5 wt % of Nb, 1.0 wt % of Sn, 0.15 wt % of Fe, and 1210 ppm of O is prepared as a 9 mm thick cold-rolled plate by a method which comprises heating at 940° C. for thirty minutes, water cooling, cold-working with a extrusion percentage of 40%, and annealing at 400° C. for twenty minutes. The alloy exhibits similar mechanical properties as that of the zirconium alloy comprising 2.5 wt % of Nb and 1230 ppm of O.
As described above with respect to the prior art, research has been carried out on conventional zirconium-based alloy containing niobium for use in nuclear fuel cladding tubes, to impart high corrosion resistance by changing the type and amount of elements to be added.